JP2003215287A - Reprocessing method for spent nuclear fuel using eutectic phenomenon - Google Patents
Reprocessing method for spent nuclear fuel using eutectic phenomenonInfo
- Publication number
- JP2003215287A JP2003215287A JP2002008320A JP2002008320A JP2003215287A JP 2003215287 A JP2003215287 A JP 2003215287A JP 2002008320 A JP2002008320 A JP 2002008320A JP 2002008320 A JP2002008320 A JP 2002008320A JP 2003215287 A JP2003215287 A JP 2003215287A
- Authority
- JP
- Japan
- Prior art keywords
- solution
- crystals
- plutonium
- mother liquor
- nuclear fuel
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Granted
Links
- 239000002915 spent fuel radioactive waste Substances 0.000 title claims abstract description 44
- 238000009376 nuclear reprocessing Methods 0.000 title claims abstract description 23
- 230000005496 eutectics Effects 0.000 title description 2
- 239000012452 mother liquor Substances 0.000 claims abstract description 55
- GRYLNZFGIOXLOG-UHFFFAOYSA-N nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims abstract description 50
- 229910052770 Uranium Inorganic materials 0.000 claims abstract description 48
- 229910052778 Plutonium Inorganic materials 0.000 claims abstract description 45
- 239000002253 acid Substances 0.000 claims abstract description 39
- QWDZADMNIUIMTC-UHFFFAOYSA-N Uranyl nitrate Chemical compound [O-][N+](=O)O[U-2](=O)(=O)O[N+]([O-])=O QWDZADMNIUIMTC-UHFFFAOYSA-N 0.000 claims abstract description 30
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 claims abstract description 29
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims abstract description 28
- -1 uranium-plutonium Chemical compound 0.000 claims abstract description 17
- 238000001816 cooling Methods 0.000 claims description 21
- 238000000034 method Methods 0.000 claims description 21
- 239000007788 liquid Substances 0.000 abstract description 11
- 239000000446 fuel Substances 0.000 abstract description 9
- 230000035755 proliferation Effects 0.000 abstract description 7
- 239000002699 waste material Substances 0.000 abstract description 7
- 230000002708 enhancing Effects 0.000 abstract description 3
- 238000002288 cocrystallisation Methods 0.000 abstract 1
- 239000000047 product Substances 0.000 description 30
- 238000005406 washing Methods 0.000 description 11
- 238000010008 shearing Methods 0.000 description 8
- STCOOQWBFONSKY-UHFFFAOYSA-N Tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 6
- 239000003758 nuclear fuel Substances 0.000 description 6
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 6
- NHNBFGGVMKEFGY-UHFFFAOYSA-N nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 description 5
- 229910002089 NOx Inorganic materials 0.000 description 4
- 238000007664 blowing Methods 0.000 description 4
- 238000009835 boiling Methods 0.000 description 4
- 239000003153 chemical reaction reagent Substances 0.000 description 4
- 238000002425 crystallisation Methods 0.000 description 4
- 230000005712 crystallization Effects 0.000 description 4
- OSFGNZOUZOPXBL-UHFFFAOYSA-N nitric acid;trihydrate Chemical compound O.O.O.O[N+]([O-])=O OSFGNZOUZOPXBL-UHFFFAOYSA-N 0.000 description 4
- 239000003960 organic solvent Substances 0.000 description 4
- 238000011084 recovery Methods 0.000 description 4
- FAPWRFPIFSIZLT-UHFFFAOYSA-M sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 description 4
- 239000002904 solvent Substances 0.000 description 4
- 239000002244 precipitate Substances 0.000 description 3
- 239000008346 aqueous phase Substances 0.000 description 2
- 238000005260 corrosion Methods 0.000 description 2
- 238000010586 diagram Methods 0.000 description 2
- 238000004090 dissolution Methods 0.000 description 2
- 238000005868 electrolysis reaction Methods 0.000 description 2
- 239000002360 explosive Substances 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 239000000463 material Substances 0.000 description 2
- 150000002894 organic compounds Chemical class 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- 125000005289 uranyl group Chemical group 0.000 description 2
- 230000001376 precipitating Effects 0.000 description 1
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
- G21C19/42—Reprocessing of irradiated fuel
- G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
- G21C19/46—Aqueous processes, e.g. by using organic extraction means, including the regeneration of these means
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Abstract
(57)【要約】
【課題】 廃液発生量を低減でき、レッドオイルが生成
せず、しかも核不拡散抵抗性を強化できるようにする。
【解決手段】 ６価のＵとＰｕの共晶析現象を利用し
て、使用済核燃料の硝酸溶液からＵ及びＵ−ＰｕとＦＰ
やＴＲＵ等を粗分離する使用済核燃料の再処理方法であ
る。例えば、使用済核燃料を剪断し、硝酸に溶解し、溶
液中の不溶解残渣を除去する。次に酸濃度及び溶液中の
プルトニウムの原子価を４価に調整し、溶液を冷却して
硝酸ウラニル結晶を析出させ、母液と結晶を分離し、結
晶をウラン製品として回収する。そして、分離した母液
について、酸濃度及び溶液中のウランとプルトニウムの
原子価を６価に調整し、溶液を冷却して硝酸ウラニル・
プルトニウム結晶を析出させ、母液と結晶を分離し、分
離した結晶をウラン・プルトニウム混合製品として回収
する。
(57) [Summary] [PROBLEMS] To reduce the amount of waste liquid generated, to prevent generation of red oil, and to enhance nuclear non-proliferation resistance. SOLUTION: U, U-Pu and FP are obtained from a nitric acid solution of spent nuclear fuel by utilizing a co-crystallization phenomenon of hexavalent U and Pu.
This is a method for reprocessing spent nuclear fuel that roughly separates fuel and TRU. For example, spent nuclear fuel is sheared and dissolved in nitric acid to remove insoluble residues in the solution. Next, the acid concentration and the valence of plutonium in the solution are adjusted to tetravalent, the solution is cooled to precipitate uranyl nitrate crystals, the mother liquor and the crystals are separated, and the crystals are recovered as a uranium product. Then, with respect to the separated mother liquor, the acid concentration and the valence of uranium and plutonium in the solution are adjusted to hexavalent, and the solution is cooled to obtain uranyl nitrate.
The plutonium crystal is precipitated, the mother liquor and the crystal are separated, and the separated crystal is recovered as a uranium-plutonium mixed product.
Description
【０００１】[0001]
【発明の属する技術分野】本発明は、ウラン・プルトニ
ウムの共晶析現象を利用して、使用済核燃料からウラン
・プルトニウム等の核燃料物質を回収する方法に関する
ものである。TECHNICAL FIELD The present invention relates to a method for recovering a nuclear fuel material such as uranium / plutonium from a spent nuclear fuel by utilizing a uranium / plutonium eutecticization phenomenon.
【０００２】[0002]
【従来の技術】使用済核燃料には、ウラン（Ｕ）やプル
トニウム（Ｐｕ）等、再び燃料として使用できる物質が
含まれている。従って、それらを回収し再利用すること
によりエネルギー資源の有効利用を図ることができる。2. Description of the Related Art Spent nuclear fuel contains substances such as uranium (U) and plutonium (Pu) that can be reused as fuel. Therefore, energy resources can be effectively used by recovering and reusing them.
【０００３】現在、実用化されている使用済核燃料の再
処理方法としては、ピューレックス（ＰＵＲＥＸ）法が
ある。この方法では、まず燃料集合体を剪断し、硝酸に
溶解する。次に、硝酸溶液を清澄し酸濃度等の調整を行
い、リン酸トリブチル（ＴＢＰ）を溶媒としてＵとＰｕ
を抽出する。更に試薬を用いて価数調整等を行い、それ
ぞれを水相に逆抽出することで核分裂生成物（ＦＰ）や
超ウラン元素（ＴＲＵ）、腐食生成物（ＣＰ）等からＵ
やＰｕを選択的に分離回収する。As a reprocessing method of spent nuclear fuel which has been put into practical use at present, there is a PUREX method. In this method, the fuel assembly is first sheared and dissolved in nitric acid. Next, the nitric acid solution is clarified to adjust the acid concentration and the like, and U and Pu are used with tributyl phosphate (TBP) as a solvent.
To extract. Furthermore, the valences are adjusted using reagents, and by back-extracting each into the aqueous phase, fission products (FP), transuranium elements (TRU), corrosion products (CP), etc.
And Pu are selectively separated and recovered.
【０００４】[0004]
【発明が解決しようとする課題】このようなＰＵＲＥＸ
法は、Ｕ及びＰｕの選択的回収、臨界管理、安全性、ハ
ンドリング等の面で優れた再処理方法である。しかし、
環境的負荷低減の観点から、工程の簡素化及び廃液発生
量の低減が求められている。また、硝酸の存在下での有
機溶媒の使用は、爆発性有機化合物（レッドオイル）が
生成してしまう可能性があり、補助的な操作並びに管理
が必要となる。更に、Ｐｕが単独で存在する工程がある
ことから、核不拡散という点に関しては若干不利といえ
る。Such a PUREX
The method is an excellent reprocessing method in terms of selective recovery of U and Pu, criticality control, safety, handling and the like. But,
From the viewpoint of reducing environmental load, simplification of processes and reduction of waste liquid generation are required. In addition, the use of an organic solvent in the presence of nitric acid may generate an explosive organic compound (red oil), which requires auxiliary operation and management. Further, since there is a process in which Pu exists alone, it can be said that it is slightly disadvantageous in terms of nuclear non-proliferation.
【０００５】本発明の目的は、廃液発生量を低減でき、
レッドオイルが生成せず、しかも核不拡散抵抗性を強化
できる使用済核燃料の再処理方法を提供することであ
る。The object of the present invention is to reduce the amount of waste liquid generated,
It is an object of the present invention to provide a method for reprocessing spent nuclear fuel that does not produce red oil and can enhance nuclear non-proliferation resistance.
【０００６】[0006]
【課題を解決するための手段】本発明は、６価のＵとＰ
ｕの共晶析現象を利用して、使用済核燃料の硝酸溶液か
らＵ及びＵ−ＰｕとＦＰやＴＲＵ等を粗分離する方法で
ある。The present invention is based on hexavalent U and P.
This is a method of roughly separating U and U-Pu from FP, TRU, etc. from a nitric acid solution of spent nuclear fuel by utilizing the eutecticization phenomenon of u.
【０００７】本発明は、使用済核燃料を剪断し、硝酸に
溶解し、溶液中の不溶解残渣を除去する工程、酸濃度及
び溶液中のウランとプルトニウムの原子価を６価に調整
し、溶液を冷却して硝酸ウラニル・プルトニウム結晶を
析出させ、母液と結晶を分離する工程、その結晶を再度
硝酸に溶解し、プルトニウムの原子価を４価に調整し、
溶液を冷却して硝酸ウラニル結晶を析出させ、母液と結
晶を分離して、結晶をウラン製品として回収し、母液を
ウラン・プルトニウム混合製品として回収する工程、を
具備している共晶析現象を利用した使用済核燃料の再処
理方法である。The present invention comprises the steps of shearing spent nuclear fuel, dissolving it in nitric acid, and removing insoluble residues in the solution, adjusting the acid concentration and the valence of uranium and plutonium in the solution to 6 Is cooled to precipitate a uranyl nitrate plutonium crystal, the mother liquor and the crystal are separated, the crystal is dissolved again in nitric acid, and the valence of plutonium is adjusted to tetravalent,
The solution is cooled to precipitate uranyl nitrate crystals, the mother liquor and crystals are separated, the crystals are recovered as a uranium product, and the mother liquor is recovered as a uranium-plutonium mixed product. This is the method of reprocessing spent nuclear fuel.
【０００８】また本発明は、使用済核燃料を剪断し、硝
酸に溶解し、溶液中の不溶解残渣を除去する工程、酸濃
度及び溶液中のプルトニウムの原子価を４価に調整し、
溶液を冷却して硝酸ウラニル結晶を析出させ、母液と結
晶を分離し、結晶をウラン製品として回収する工程、分
離した母液について、酸濃度及び溶液中のウランとプル
トニウムの原子価を６価に調整し、溶液を冷却して硝酸
ウラニル・プルトニウム結晶を析出させ、母液と結晶を
分離し、分離した結晶をウラン・プルトニウム混合製品
として回収する工程、を具備している共晶析現象を利用
した使用済核燃料の再処理方法である。The present invention further comprises the steps of shearing spent nuclear fuel, dissolving it in nitric acid, and removing insoluble residues in the solution, adjusting the acid concentration and the valence of plutonium in the solution to be tetravalent,
Cooling the solution to precipitate uranyl nitrate crystals, separating the mother liquor and the crystals, and recovering the crystals as uranium products. Adjusting the acid concentration and valence of uranium and plutonium in the solution to hexavalent for the separated mother liquor. Then, the solution is cooled to precipitate the uranyl-plutonium nitrate crystal, the mother liquor and the crystal are separated, and the separated crystal is recovered as a uranium-plutonium mixed product. It is a method of reprocessing nuclear fuel.
【０００９】[0009]
【実施例】図１は、本発明に係る共晶析現象を利用した
使用済核燃料の再処理方法の一実施例を示す工程説明図
である。EXAMPLE FIG. 1 is a process diagram showing an example of a method for reprocessing spent nuclear fuel utilizing the eutecticization phenomenon according to the present invention.
【００１０】まず工程Ａでは、使用済核燃料を剪断し、
硝酸に溶解し、溶液中の不溶解残渣を除去する。更に詳
しく述べると、
（ａ）使用済核燃料（燃料ピン束）を剪断機で小片に剪
断する。
（ｂ）剪断片を硝酸で溶解する。これには、例えば１３
Ｎ以下、１１０℃程度の沸騰硝酸を用いる。
（ｃ）そして、清澄した溶液を次工程に移送する。これ
によって、溶液中の不溶解残渣が除去される。First, in step A, the spent nuclear fuel is sheared,
Dissolve in nitric acid and remove the insoluble residue in the solution. More specifically, (a) the spent nuclear fuel (fuel pin bundle) is sheared into small pieces by a shearing machine. (B) Dissolve the shear pieces with nitric acid. This includes, for example, 13
Boiling nitric acid having a temperature of N or less and about 110 ° C. is used. (C) Then, the clarified solution is transferred to the next step. This removes insoluble residues in the solution.
【００１１】次に工程Ｂでは、酸濃度及び溶液中のウラ
ンとプルトニウムの原子価を６価に調整し、溶液を冷却
して硝酸ウラニル・プルトニウム結晶を析出させ、母液
と結晶を分離し、分離した結晶を洗浄する。これによっ
て、ＦＰ，ＴＲＵ，ＣＰ等が取り除かれる。更に詳しく
述べると、
（ｄ）溶液の酸濃度を６Ｍ程度に調整し、溶液中のＵ，
Ｐｕの原子価を電解により４価から６価に調整する。
（ｅ）溶液を冷却することにより、硝酸ウラニル（ＵＮ
Ｈ）及び硝酸ウラニル・プルトニル（（Ｕ，Ｐｕ）Ｎ
Ｈ）結晶を析出させ、母液とそれらの結晶を分離する。
（ｆ）得られた結晶を、母液と同程度の酸濃度の硝酸で
洗浄する。Next, in step B, the acid concentration and the valences of uranium and plutonium in the solution are adjusted to 6 and the solution is cooled to precipitate uranyl-plutonium nitrate crystals, and the mother liquor and the crystals are separated and separated. The washed crystals are washed. As a result, FP, TRU, CP, etc. are removed. More specifically, (d) adjusting the acid concentration of the solution to about 6 M,
The valence of Pu is adjusted from 4 to 6 by electrolysis. (E) By cooling the solution, uranyl nitrate (UN
H) and uranyl plutonyl nitrate ((U, Pu) N
H) Precipitate crystals and separate the mother liquor and those crystals. (F) The obtained crystals are washed with nitric acid having an acid concentration similar to that of the mother liquor.
【００１２】更に工程Ｃでは、それらの結晶を再度硝酸
に溶解し、プルトニウムの原子価を４価に調整し、溶液
を冷却して硝酸ウラニル結晶を析出させ、母液と結晶を
分離し、結晶を洗浄して、該結晶をウラン製品として回
収し、母液をウラン・プルトニウム混合製品として回収
する。詳しく述べると、
（ｇ）得られたＵＮＨ結晶及び（Ｕ，Ｐｕ）ＮＨ結晶を
再度硝酸に溶解する。これらの結晶は水及び希硝酸に易
溶性であり、濃度４Ｍ程度で温度４０℃程度の硝酸で溶
解する。
（ｈ）溶液中のＰｕの原子価を４価に調整する。この調
整は、ＮＯｘガスの吹き込みによって行う。また、次の
ステップ（ｉ）の冷却過程において水（Ｈ2 Ｏ）や硝酸
３水和物（ＨＮＯ3 ・３Ｈ2 Ｏ）の結晶が生成しないよ
うに、酸濃度の調整を行う。酸濃度は６Ｍ程度とする。
（ｉ）溶液を４０℃から１０℃程度まで冷却することに
より、ＵＮＨ結晶を析出させ、母液と結晶を分離する。
（ｊ）結晶を母液と同程度の酸濃度の硝酸で洗浄し、結
晶をＵ製品、母液及び洗浄液をＵ−Ｐｕ混合製品として
回収する。Further, in step C, the crystals are again dissolved in nitric acid, the valence of plutonium is adjusted to tetravalent, the solution is cooled to precipitate uranyl nitrate crystals, and the mother liquor and the crystals are separated to form crystals. After washing, the crystals are recovered as a uranium product, and the mother liquor is recovered as a uranium-plutonium mixed product. More specifically, (g) the obtained UNH crystal and (U, Pu) NH crystal are dissolved again in nitric acid. These crystals are easily soluble in water and dilute nitric acid, and are dissolved in nitric acid having a concentration of about 4M and a temperature of about 40 ° C. (H) The valence of Pu in the solution is adjusted to be tetravalent. This adjustment is performed by blowing NOx gas. Further, the acid concentration is adjusted so that crystals of water (H 2 O) or nitric acid trihydrate (HNO 3 .3H 2 O) are not generated in the cooling process of the next step (i). The acid concentration is about 6M. (I) By cooling the solution from 40 ° C. to about 10 ° C., UNH crystals are precipitated and the mother liquor and the crystals are separated. (J) The crystals are washed with nitric acid having the same acid concentration as that of the mother liquor, and the crystals are collected as a U product, and the mother liquor and the washing liquid are collected as a U-Pu mixed product.
【００１３】図２は、本発明に係る共晶析現象を利用し
た使用済核燃料の再処理方法の他の実施例を示す工程説
明図である。FIG. 2 is a process explanatory view showing another embodiment of the spent nuclear fuel reprocessing method utilizing the eutecticization phenomenon according to the present invention.
【００１４】まず工程Ｄでは、使用済核燃料を剪断し、
硝酸に溶解し、溶液中の不溶解残渣を除去する。この工
程は、前記実施例と同様である。詳しく述べると、
（ｋ）使用済核燃料（燃料ピン束）を剪断機で小片に剪
断する。
（ｌ）剪断片を硝酸で溶解する。これには、例えば１３
Ｎ以下、１１０℃程度の沸騰硝酸を用いる。
（ｍ）そして、清澄した溶液を次工程に移送する。これ
によって、溶液中の不溶解残渣が除去される。First, in step D, the spent nuclear fuel is sheared,
Dissolve in nitric acid and remove the insoluble residue in the solution. This step is the same as in the above embodiment. Specifically, (k) the spent nuclear fuel (fuel pin bundle) is sheared into small pieces by a shearing machine. (L) Dissolve the shear pieces with nitric acid. This includes, for example, 13
Boiling nitric acid having a temperature of N or less and about 110 ° C. is used. (M) Then, the clarified solution is transferred to the next step. This removes insoluble residues in the solution.
【００１５】次に工程Ｅでは、酸濃度及び溶液中のプル
トニウムの原子価を４価に調整し、溶液を冷却して硝酸
ウラニル結晶を析出させ、母液と結晶を分離し、結晶を
洗浄してウラン製品として回収する。更に詳しく述べる
と、
（ｎ）溶液の酸濃度を６Ｍ程度に調整し、溶液中のＰｕ
の原子価を４価から６価に調整する。この原子価の調整
は、ＮＯｘガスの吹き込みにより行う。酸濃度の調整
は、次のステップ（ｏ）の冷却過程において水（Ｈ
2 Ｏ）や硝酸３水和物（ＨＮＯ3 ・３Ｈ2 Ｏ）の結晶が
生成しないようにするためである。
（ｏ）溶液を４０℃から１０℃程度まで冷却することに
より、ＵＮＨ結晶を析出させ、母液と結晶を分離する。
（ｐ）結晶を母液と同程度の酸濃度の硝酸で洗浄し、結
晶をＵ製品として回収する。Next, in step E, the acid concentration and the valence of plutonium in the solution are adjusted to be tetravalent, the solution is cooled to precipitate uranyl nitrate crystals, the mother liquor and the crystals are separated, and the crystals are washed. Recover as uranium product. More specifically, (n) the acid concentration of the solution was adjusted to about 6M, and the Pu in the solution was adjusted.
The valence of is adjusted from 4 to 6. The adjustment of the valence is performed by blowing NOx gas. The acid concentration can be adjusted by adding water (H) in the cooling process of the next step (o).
2 O) and nitric acid trihydrate (HNO 3 · 3H 2 O) crystals are not generated. (O) By cooling the solution from 40 ° C. to about 10 ° C., UNH crystals are precipitated and the mother liquor and the crystals are separated. (P) The crystals are washed with nitric acid having an acid concentration similar to that of the mother liquor, and the crystals are recovered as a U product.
【００１６】更に工程Ｆでは、分離した母液について、
酸濃度及び溶液中のウランとプルトニウムの原子価を６
価に調整し、溶液を冷却して硝酸ウラニル・プルトニウ
ム結晶を析出させ、母液と結晶を分離して、分離した結
晶を洗浄し、結晶をウラン・プルトニウム混合製品とし
て回収する。その際、ＦＰ，ＴＲＵ，ＣＰ等が取り除か
れる。更に詳しく述べると、
（ｑ）分離した母液について、酸濃度及び溶液中のＵ，
Ｐｕの原子価を６価に調整する。
（ｒ）溶液を４０℃から−３０℃程度に冷却することに
より、硝酸ウラニル・プルトニル（（Ｕ，Ｐｕ）ＮＨ）
結晶を析出させ、母液と該結晶を分離する。
（ｓ）母液と同程度の酸濃度の硝酸で結晶を洗浄し、結
晶をＵ−Ｐｕ混合製品として回収する。In step F, the separated mother liquor is
The acid concentration and the valence of uranium and plutonium in the solution were determined to be 6
The valence is adjusted, the solution is cooled to precipitate uranyl-plutonium nitrate crystals, the mother liquor and crystals are separated, the separated crystals are washed, and the crystals are recovered as a uranium-plutonium mixed product. At this time, FP, TRU, CP, etc. are removed. More specifically, (q) with respect to the separated mother liquor, the acid concentration and U in the solution,
The valence of Pu is adjusted to be hexavalent. (R) By cooling the solution from 40 ° C. to about −30 ° C., uranyl-plutonyl nitrate ((U, Pu) NH)
Crystals are precipitated and the mother liquor and the crystals are separated. (S) The crystals are washed with nitric acid having the same acid concentration as that of the mother liquor, and the crystals are collected as a U-Pu mixed product.
【００１７】上記の各実施例において、晶析や洗浄の過
程で発生する母液や洗浄液の全部あるいは一部を蒸留す
ること等によって硝酸を分離すると、溶解、調整、又は
洗浄に再利用することができる。In each of the above examples, when the nitric acid is separated by distilling all or part of the mother liquor and the washing liquid generated in the process of crystallization and washing, it can be reused for dissolution, adjustment, or washing. it can.
【００１８】ステップ（ｉ）又は（ｏ）において、晶析
温度や時間によって回収するＵ量を調節することによ
り、最終的に回収されるＵ−Ｐｕ混合製品のＵ／Ｐｕ比
を調整することができる。Ｕの溶解度曲線（硝酸溶液濃
度をパラメータとしたときの溶液温度に対するＵ濃度）
は既知であり、従ってその溶解度曲線から収率を冷却温
度や濃度調整で制御できるからである。例えば、上記実
施例のように、４０℃から１０℃に冷却することによ
り、Ｕのおよそ６０％が回収でき、Ｕ／Ｐｕは３：１に
なる。In step (i) or (o), the U / Pu ratio of the U-Pu mixed product finally recovered can be adjusted by adjusting the amount of U recovered depending on the crystallization temperature and time. it can. Solubility curve of U (U concentration against solution temperature when nitric acid solution concentration is used as a parameter)
Is known and therefore the yield can be controlled from the solubility curve by adjusting the cooling temperature and the concentration. For example, by cooling from 40 ° C. to 10 ° C. as in the above example, about 60% of U can be recovered and U / Pu becomes 3: 1.
【００１９】[0019]
【発明の効果】本発明は上記のように、ウラン・プルト
ニウムの共晶析現象を利用して、使用済核燃料からＵと
Ｐｕを回収する方法であるから、従来のＰＵＲＥＸ法と
比較して工程中で使用する試薬や溶媒の量を減らすこと
ができ、廃液発生量の低減を図ることができる。また、
ＴＢＰ等の有機溶媒を使用しないことから、可燃性のレ
ッドオイルが生成することもない。更に、ＵとＰｕを同
時に回収し、Ｐｕ単体での回収を行わないために、核不
拡散抵抗性を強化することができる。As described above, the present invention is a method for recovering U and Pu from spent nuclear fuel by utilizing the uranium-plutonium eutecticization phenomenon. The amount of reagents and solvents used therein can be reduced, and the amount of waste liquid generated can be reduced. Also,
Since no organic solvent such as TBP is used, flammable red oil is not produced. Furthermore, since U and Pu are recovered at the same time and the recovery of Pu alone is not performed, the nuclear non-proliferation resistance can be enhanced.
【図１】本発明に係る使用済核燃料の再処理方法の一実
施例を示す工程説明図。FIG. 1 is a process explanatory view showing an embodiment of a spent nuclear fuel reprocessing method according to the present invention.
【図２】本発明に係る使用済核燃料の再処理方法の他の
実施例を示す工程説明図。FIG. 2 is a process explanatory view showing another embodiment of the spent nuclear fuel reprocessing method according to the present invention.
─────────────────────────────────────────────────────
─────────────────────────────────────────────────── ───
【手続補正書】[Procedure amendment]
【提出日】平成１４年８月２８日（２００２．８．２
８）[Submission date] August 28, 2002 (2002.8.2)
8)
【手続補正１】[Procedure Amendment 1]
【補正対象書類名】明細書[Document name to be amended] Statement
【補正対象項目名】全文[Correction target item name] Full text
【補正方法】変更[Correction method] Change
【補正内容】[Correction content]
【書類名】 明細書[Document name] Statement
【発明の名称】 共晶析現象を利用した使用済核燃料の
再処理方法Title of the invention Reprocessing method of spent nuclear fuel utilizing eutectic phenomenon
【特許請求の範囲】[Claims]
【発明の詳細な説明】Detailed Description of the Invention
【０００１】[0001]
【発明の属する技術分野】本発明は、ウラン・プルトニ
ウムの共晶析現象を利用して、使用済核燃料からウラン
・プルトニウム等の核燃料物質を回収する方法に関する
ものである。TECHNICAL FIELD The present invention relates to a method for recovering a nuclear fuel material such as uranium / plutonium from a spent nuclear fuel by utilizing a uranium / plutonium eutecticization phenomenon.
【０００２】[0002]
【従来の技術】使用済核燃料には、ウラン（Ｕ）やプル
トニウム（Ｐｕ）等、再び燃料として使用できる物質が
含まれている。従って、それらを回収し再利用すること
によりエネルギー資源の有効利用を図ることができる。2. Description of the Related Art Spent nuclear fuel contains substances such as uranium (U) and plutonium (Pu) that can be reused as fuel. Therefore, energy resources can be effectively used by recovering and reusing them.
【０００３】現在、実用化されている使用済核燃料の再
処理方法としては、ピューレックス（ＰＵＲＥＸ）法が
ある。この方法では、まず燃料集合体を剪断し、硝酸に
溶解する。次に、硝酸溶液を清澄し酸濃度等の調整を行
い、リン酸トリブチル（ＴＢＰ）を溶媒としてＵとＰｕ
を抽出する。更に試薬を用いて価数調整等を行い、それ
ぞれを水相に逆抽出することで核分裂生成物（ＦＰ）や
超ウラン元素（ＴＲＵ）、腐食生成物（ＣＰ）等からＵ
やＰｕを選択的に分離回収する。As a reprocessing method of spent nuclear fuel which has been put into practical use at present, there is a PUREX method. In this method, the fuel assembly is first sheared and dissolved in nitric acid. Next, the nitric acid solution is clarified to adjust the acid concentration and the like, and U and Pu are used with tributyl phosphate (TBP) as a solvent.
To extract. Furthermore, the valences are adjusted using reagents, and by back-extracting each into the aqueous phase, fission products (FP), transuranium elements (TRU), corrosion products (CP), etc.
And Pu are selectively separated and recovered.
【０００４】[0004]
【発明が解決しようとする課題】このようなＰＵＲＥＸ
法は、Ｕ及びＰｕの選択的回収、臨界管理、安全性、ハ
ンドリング等の面で優れた再処理方法である。しかし、
環境的負荷低減の観点から、工程の簡素化及び廃液発生
量の低減が求められている。また、硝酸の存在下での有
機溶媒の使用は、爆発性有機化合物（レッドオイル）が
生成してしまう可能性があり、補助的な操作並びに管理
が必要となる。更に、Ｐｕが単独で存在する工程がある
ことから、核不拡散という点に関しては若干不利といえ
る。Such a PUREX
The method is an excellent reprocessing method in terms of selective recovery of U and Pu, criticality control, safety, handling and the like. But,
From the viewpoint of reducing environmental load, simplification of processes and reduction of waste liquid generation are required. In addition, the use of an organic solvent in the presence of nitric acid may generate an explosive organic compound (red oil), which requires auxiliary operation and management. Further, since there is a process in which Pu exists alone, it can be said that it is slightly disadvantageous in terms of nuclear non-proliferation.
【０００５】本発明の目的は、廃液発生量を低減でき、
レッドオイルが生成せず、しかも核不拡散抵抗性を強化
できる使用済核燃料の再処理方法を提供することであ
る。The object of the present invention is to reduce the amount of waste liquid generated,
It is an object of the present invention to provide a method for reprocessing spent nuclear fuel that does not produce red oil and can enhance nuclear non-proliferation resistance.
【０００６】[0006]
【課題を解決するための手段】本発明は、６価のＵとＰ
ｕの共晶析現象を利用して、使用済核燃料の硝酸溶液か
らＵ及びＵ−ＰｕとＦＰやＴＲＵ等を粗分離する方法で
ある。The present invention is based on hexavalent U and P.
This is a method of roughly separating U and U-Pu from FP, TRU, etc. from a nitric acid solution of spent nuclear fuel by utilizing the eutecticization phenomenon of u.
【０００７】本発明は、使用済核燃料を剪断し、硝酸に
溶解し、溶液中の不溶解残渣を除去する工程、酸濃度及
び溶液中のウランとプルトニウムの原子価を６価に調整
し、溶液を冷却して硝酸ウラニル・プルトニル結晶を析
出させ、母液と結晶を分離する工程、その結晶を再度硝
酸に溶解し、プルトニウムの原子価を４価に調整し、溶
液を冷却して硝酸ウラニル結晶を析出させ、母液と結晶
を分離して、結晶をウラン製品として回収し、母液をウ
ラン・プルトニウム混合製品として回収する工程、を具
備している共晶析現象を利用した使用済核燃料の再処理
方法である。The present invention comprises the steps of shearing spent nuclear fuel, dissolving it in nitric acid, and removing insoluble residues in the solution, adjusting the acid concentration and the valence of uranium and plutonium in the solution to 6 the precipitate was cooled uranyl nitrate-plutonium Le crystals, separating the crystals with the mother liquor and dissolved in the crystal again nitrate to adjust the valence of plutonium tetravalent, the solution was cooled uranyl nitrate crystals Reprocessing of spent nuclear fuel utilizing the eutecticization phenomenon, including the steps of precipitating and separating the mother liquor and crystals, collecting the crystals as a uranium product, and collecting the mother liquor as a uranium-plutonium mixed product. Is the way.
【０００８】また本発明は、使用済核燃料を剪断し、硝
酸に溶解し、溶液中の不溶解残渣を除去する工程、酸濃
度及び溶液中のプルトニウムの原子価を４価に調整し、
溶液を冷却して硝酸ウラニル結晶を析出させ、母液と結
晶を分離し、結晶をウラン製品として回収する工程、分
離した母液について、酸濃度及び溶液中のウランとプル
トニウムの原子価を６価に調整し、溶液を冷却して硝酸
ウラニル・プルトニル結晶を析出させ、母液と結晶を分
離し、分離した結晶をウラン・プルトニウム混合製品と
して回収する工程、を具備している共晶析現象を利用し
た使用済核燃料の再処理方法である。The present invention further comprises the steps of shearing spent nuclear fuel, dissolving it in nitric acid, and removing insoluble residues in the solution, adjusting the acid concentration and the valence of plutonium in the solution to be tetravalent,
Cooling the solution to precipitate uranyl nitrate crystals, separating the mother liquor and the crystals, and recovering the crystals as uranium products. Adjusting the acid concentration and valence of uranium and plutonium in the solution to hexavalent for the separated mother liquor. and the solution allowed to precipitate by cooling uranyl nitrate-plutonium Le crystals, the crystals with the mother liquor were separated and the separated crystals using KyoAkira析behavior that comprises steps, the recovered as uranium-plutonium mixed product This is a method of reprocessing spent nuclear fuel.
【０００９】[0009]
【実施例】図１は、本発明に係る共晶析現象を利用した
使用済核燃料の再処理方法の一実施例を示す工程説明図
である。EXAMPLE FIG. 1 is a process diagram showing an example of a method for reprocessing spent nuclear fuel utilizing the eutecticization phenomenon according to the present invention.
【００１０】まず工程Ａでは、使用済核燃料を剪断し、
硝酸に溶解し、溶液中の不溶解残渣を除去する。更に詳
しく述べると、
（ａ）使用済核燃料（燃料ピン束）を剪断機で小片に剪
断する。
（ｂ）剪断片を硝酸で溶解する。これには、例えば１３
Ｎ以下、１１０℃程度の沸騰硝酸を用いる。
（ｃ）そして、清澄した溶液を次工程に移送する。これ
によって、溶液中の不溶解残渣が除去される。First, in step A, the spent nuclear fuel is sheared,
Dissolve in nitric acid and remove the insoluble residue in the solution. More specifically, (a) the spent nuclear fuel (fuel pin bundle) is sheared into small pieces by a shearing machine. (B) Dissolve the shear pieces with nitric acid. This includes, for example, 13
Boiling nitric acid having a temperature of N or less and about 110 ° C. is used. (C) Then, the clarified solution is transferred to the next step. This removes insoluble residues in the solution.
【００１１】次に工程Ｂでは、酸濃度及び溶液中のウラ
ンとプルトニウムの原子価を６価に調整し、溶液を冷却
して硝酸ウラニル・プルトニル結晶を析出させ、母液と
結晶を分離し、分離した結晶を洗浄する。これによっ
て、ＦＰ，ＴＲＵ，ＣＰ等が取り除かれる。更に詳しく
述べると、
（ｄ）溶液の酸濃度を６Ｍ程度に調整し、溶液中のＵ，
Ｐｕの原子価を電解により４価から６価に調整する。
（ｅ）溶液を冷却することにより、硝酸ウラニル（ＵＮ
Ｈ）及び硝酸ウラニル・プルトニル（（Ｕ，Ｐｕ）Ｎ
Ｈ）結晶を析出させ、母液とそれらの結晶を分離する。
（ｆ）得られた結晶を、母液と同程度の酸濃度の硝酸で
洗浄する。[0011] In the next step B, a valence uranium and plutonium acid concentration and solution was adjusted to hexavalent, solution was cooled to precipitate the uranyl nitrate-plutonium Le crystals, the crystals were separated with the mother liquor, The separated crystals are washed. As a result, FP, TRU, CP, etc. are removed. More specifically, (d) adjusting the acid concentration of the solution to about 6 M,
The valence of Pu is adjusted from 4 to 6 by electrolysis. (E) By cooling the solution, uranyl nitrate (UN
H) and uranyl plutonyl nitrate ((U, Pu) N
H) Precipitate crystals and separate the mother liquor and those crystals. (F) The obtained crystals are washed with nitric acid having an acid concentration similar to that of the mother liquor.
【００１２】更に工程Ｃでは、それらの結晶を再度硝酸
に溶解し、プルトニウムの原子価を４価に調整し、溶液
を冷却して硝酸ウラニル結晶を析出させ、母液と結晶を
分離し、結晶を洗浄して、該結晶をウラン製品として回
収し、母液をウラン・プルトニウム混合製品として回収
する。詳しく述べると、
（ｇ）得られたＵＮＨ結晶及び（Ｕ，Ｐｕ）ＮＨ結晶を
再度硝酸に溶解する。これらの結晶は水及び希硝酸に易
溶性であり、濃度４Ｍ程度で温度４０℃程度の硝酸で溶
解する。
（ｈ）溶液中のＰｕの原子価を４価に調整する。この調
整は、ＮＯｘガスの吹き込みによって行う。また、次の
ステップ（ｉ）の冷却過程において水（Ｈ2 Ｏ）や硝酸
３水和物（ＨＮＯ3 ・３Ｈ2 Ｏ）の結晶が生成しないよ
うに、酸濃度の調整を行う。酸濃度は６Ｍ程度とする。
（ｉ）溶液を４０℃から１０℃程度まで冷却することに
より、ＵＮＨ結晶を析出させ、母液と結晶を分離する。
（ｊ）結晶を母液と同程度の酸濃度の硝酸で洗浄し、結
晶をＵ製品、母液及び洗浄液をＵ−Ｐｕ混合製品として
回収する。Further, in step C, the crystals are again dissolved in nitric acid, the valence of plutonium is adjusted to tetravalent, the solution is cooled to precipitate uranyl nitrate crystals, and the mother liquor and the crystals are separated to form crystals. After washing, the crystals are recovered as a uranium product, and the mother liquor is recovered as a uranium-plutonium mixed product. More specifically, (g) the obtained UNH crystal and (U, Pu) NH crystal are dissolved again in nitric acid. These crystals are easily soluble in water and dilute nitric acid, and are dissolved in nitric acid having a concentration of about 4M and a temperature of about 40 ° C. (H) The valence of Pu in the solution is adjusted to be tetravalent. This adjustment is performed by blowing NOx gas. Further, the acid concentration is adjusted so that crystals of water (H 2 O) or nitric acid trihydrate (HNO 3 .3H 2 O) are not generated in the cooling process of the next step (i). The acid concentration is about 6M. (I) By cooling the solution from 40 ° C. to about 10 ° C., UNH crystals are precipitated and the mother liquor and the crystals are separated. (J) The crystals are washed with nitric acid having the same acid concentration as that of the mother liquor, and the crystals are collected as a U product, and the mother liquor and the washing liquid are collected as a U-Pu mixed product.
【００１３】図２は、本発明に係る共晶析現象を利用し
た使用済核燃料の再処理方法の他の実施例を示す工程説
明図である。FIG. 2 is a process explanatory view showing another embodiment of the spent nuclear fuel reprocessing method utilizing the eutecticization phenomenon according to the present invention.
【００１４】まず工程Ｄでは、使用済核燃料を剪断し、
硝酸に溶解し、溶液中の不溶解残渣を除去する。この工
程は、前記実施例と同様である。詳しく述べると、
（ｋ）使用済核燃料（燃料ピン束）を剪断機で小片に剪
断する。
（ｌ）剪断片を硝酸で溶解する。これには、例えば１３
Ｎ以下、１１０℃程度の沸騰硝酸を用いる。
（ｍ）そして、清澄した溶液を次工程に移送する。これ
によって、溶液中の不溶解残渣が除去される。First, in step D, the spent nuclear fuel is sheared,
Dissolve in nitric acid and remove the insoluble residue in the solution. This step is the same as in the above embodiment. Specifically, (k) the spent nuclear fuel (fuel pin bundle) is sheared into small pieces by a shearing machine. (L) Dissolve the shear pieces with nitric acid. This includes, for example, 13
Boiling nitric acid having a temperature of N or less and about 110 ° C. is used. (M) Then, the clarified solution is transferred to the next step. This removes insoluble residues in the solution.
【００１５】次に工程Ｅでは、酸濃度及び溶液中のプル
トニウムの原子価を４価に調整し、溶液を冷却して硝酸
ウラニル結晶を析出させ、母液と結晶を分離し、結晶を
洗浄してウラン製品として回収する。更に詳しく述べる
と、
（ｎ）溶液の酸濃度を６Ｍ程度に調整し、溶液中のＰｕ
の原子価を６価から４価に調整する。この原子価の調整
は、ＮＯｘガスの吹き込みにより行う。酸濃度の調整
は、次のステップ（ｏ）の冷却過程において水（Ｈ
2 Ｏ）や硝酸３水和物（ＨＮＯ3 ・３Ｈ2 Ｏ）の結晶が
生成しないようにするためである。
（ｏ）溶液を４０℃から１０℃程度まで冷却することに
より、ＵＮＨ結晶を析出させ、母液と結晶を分離する。
（ｐ）結晶を母液と同程度の酸濃度の硝酸で洗浄し、結
晶をＵ製品として回収する。Next, in step E, the acid concentration and the valence of plutonium in the solution are adjusted to be tetravalent, the solution is cooled to precipitate uranyl nitrate crystals, the mother liquor and the crystals are separated, and the crystals are washed. Recover as uranium product. More specifically, (n) the acid concentration of the solution was adjusted to about 6M, and the Pu in the solution was adjusted.
The valence of is adjusted from 6 to 4 . The adjustment of the valence is performed by blowing NOx gas. The acid concentration can be adjusted by adding water (H) in the cooling process of the next step (o).
2 O) and nitric acid trihydrate (HNO 3 · 3H 2 O) crystals are not generated. (O) By cooling the solution from 40 ° C. to about 10 ° C., UNH crystals are precipitated and the mother liquor and the crystals are separated. (P) The crystals are washed with nitric acid having an acid concentration similar to that of the mother liquor, and the crystals are recovered as a U product.
【００１６】更に工程Ｆでは、分離した母液について、
酸濃度及び溶液中のウランとプルトニウムの原子価を６
価に調整し、溶液を冷却して硝酸ウラニル・プルトニル
結晶を析出させ、母液と結晶を分離して、分離した結晶
を洗浄し、結晶をウラン・プルトニウム混合製品として
回収する。その際、ＦＰ，ＴＲＵ，ＣＰ等が取り除かれ
る。更に詳しく述べると、
（ｑ）分離した母液について、酸濃度及び溶液中のＵ，
Ｐｕの原子価を６価に調整する。
（ｒ）溶液を４０℃から−３０℃程度に冷却することに
より、硝酸ウラニル・プルトニル（（Ｕ，Ｐｕ）ＮＨ）
結晶を析出させ、母液と該結晶を分離する。
（ｓ）母液と同程度の酸濃度の硝酸で結晶を洗浄し、結
晶をＵ−Ｐｕ混合製品として回収する。In step F, the separated mother liquor is
The acid concentration and the valence of uranium and plutonium in the solution were determined to be 6
Adjust to the valence, the solution was precipitated cooled to uranyl nitrate-plutonium le <br/> crystals, and separating the crystals with the mother liquor, washing the separated crystals, recovering the crystals as mixed uranium-plutonium product. At this time, FP, TRU, CP, etc. are removed. More specifically, (q) with respect to the separated mother liquor, the acid concentration and U in the solution,
The valence of Pu is adjusted to be hexavalent. (R) By cooling the solution from 40 ° C. to about −30 ° C., uranyl-plutonyl nitrate ((U, Pu) NH)
Crystals are precipitated and the mother liquor and the crystals are separated. (S) The crystals are washed with nitric acid having the same acid concentration as that of the mother liquor, and the crystals are collected as a U-Pu mixed product.
【００１７】上記の各実施例において、晶析や洗浄の過
程で発生する母液や洗浄液の全部あるいは一部を蒸留す
ること等によって硝酸を分離すると、溶解、調整、又は
洗浄に再利用することができる。In each of the above examples, when the nitric acid is separated by distilling all or part of the mother liquor and the washing liquid generated in the process of crystallization and washing, it can be reused for dissolution, adjustment, or washing. it can.
【００１８】ステップ（ｉ）又は（ｏ）において、晶析
温度や時間によって回収するＵ量を調節することによ
り、最終的に回収されるＵ−Ｐｕ混合製品のＵ／Ｐｕ比
を調整することができる。Ｕの溶解度曲線（硝酸溶液濃
度をパラメータとしたときの溶液温度に対するＵ濃度）
は既知であり、従ってその溶解度曲線から収率を冷却温
度や濃度調整で制御できるからである。例えば、上記実
施例のように、４０℃から１０℃に冷却することによ
り、Ｕのおよそ６０％が回収でき、Ｕ／Ｐｕは３：１に
なる。In step (i) or (o), the U / Pu ratio of the U-Pu mixed product finally recovered can be adjusted by adjusting the amount of U recovered depending on the crystallization temperature and time. it can. Solubility curve of U (U concentration against solution temperature when nitric acid solution concentration is used as a parameter)
Is known and therefore the yield can be controlled from the solubility curve by adjusting the cooling temperature and the concentration. For example, by cooling from 40 ° C. to 10 ° C. as in the above example, about 60% of U can be recovered and U / Pu becomes 3: 1.
【００１９】[0019]
【発明の効果】本発明は上記のように、ウラン・プルト
ニウムの共晶析現象を利用して、使用済核燃料からＵと
Ｐｕを回収する方法であるから、従来のＰＵＲＥＸ法と
比較して工程中で使用する試薬や溶媒の量を減らすこと
ができ、廃液発生量の低減を図ることができる。また、
ＴＢＰ等の有機溶媒を使用しないことから、可燃性のレ
ッドオイルが生成することもない。更に、ＵとＰｕを同
時に回収し、Ｐｕ単体での回収を行わないために、核不
拡散抵抗性を強化することができる。As described above, the present invention is a method for recovering U and Pu from spent nuclear fuel by utilizing the uranium-plutonium eutecticization phenomenon. The amount of reagents and solvents used therein can be reduced, and the amount of waste liquid generated can be reduced. Also,
Since no organic solvent such as TBP is used, flammable red oil is not produced. Furthermore, since U and Pu are recovered at the same time and the recovery of Pu alone is not performed, the nuclear non-proliferation resistance can be enhanced.
【図面の簡単な説明】[Brief description of drawings]
【図１】本発明に係る使用済核燃料の再処理方法の一実
施例を示す工程説明図。FIG. 1 is a process explanatory view showing an embodiment of a spent nuclear fuel reprocessing method according to the present invention.
【図２】本発明に係る使用済核燃料の再処理方法の他の
実施例を示す工程説明図。FIG. 2 is a process explanatory view showing another embodiment of the spent nuclear fuel reprocessing method according to the present invention.
フロントページの続き (51)Int.Cl.7 識別記号 ＦＩ テーマコート゛(参考） Ｂ０１Ｄ 9/02 ６２５ Ｂ０１Ｄ 9/02 ６２５Ｚ (72)発明者 野村 和則 茨城県那珂郡東海村大字村松４番地33 核 燃料サイクル開発機構東海事業所内 (72)発明者 平野 弘康 茨城県那珂郡東海村大字村松４番地33 核 燃料サイクル開発機構東海事業所内 (72)発明者 青嶋 厚 茨城県那珂郡東海村大字村松４番地33 核 燃料サイクル開発機構東海事業所内Continuation of the front page (51) Int.Cl. 7 Identification code FI theme code (reference) B01D 9/02 625 B01D 9/02 625Z (72) Inventor Kazunori Nomura 4-3 Muramatsu, Tokai-mura, Naka-gun, Ibaraki Prefecture Nuclear fuel Cycle Development Organization Tokai Works (72) Inventor Hiroyasu Hirano 4-3 Muramatsu, Tokai-mura, Naka-gun, Ibaraki Prefecture 33 Nuclear Fuel Cycle Development Organization Tokai Works (72) Atsushi Aoshima 4-3, Muramatsu, Tokai-mura, Naka-gun, Ibaraki Prefecture Tokai Plant, Nuclear Fuel Cycle Development Organization
Claims (2)
溶液中の不溶解残渣を除去する工程、酸濃度及び溶液中
のウランとプルトニウムの原子価を６価に調整し、溶液
を冷却して硝酸ウラニル・プルトニウム結晶を析出さ
せ、母液と結晶を分離する工程、その結晶を再度硝酸に
溶解し、プルトニウムの原子価を４価に調整し、溶液を
冷却して硝酸ウラニル結晶を析出させ、母液と結晶を分
離して、結晶をウラン製品として回収し、母液をウラン
・プルトニウム混合製品として回収する工程、を具備し
ている共晶析現象を利用した使用済核燃料の再処理方
法。1. The spent nuclear fuel is sheared and dissolved in nitric acid,
Step of removing the insoluble residue in the solution, adjusting the acid concentration and the valence of uranium and plutonium in the solution to hexavalent, cooling the solution to precipitate the uranyl-plutonium nitrate crystal, and separating the crystal from the mother liquor Step, the crystal is again dissolved in nitric acid, the valence of plutonium is adjusted to tetravalent, the solution is cooled to precipitate the uranyl nitrate crystal, the mother liquor and the crystal are separated, the crystal is recovered as a uranium product, A process for reprocessing a spent nuclear fuel utilizing a eutecticization phenomenon, which comprises a step of recovering a mother liquor as a uranium-plutonium mixed product.
溶液中の不溶解残渣を除去する工程、酸濃度及び溶液中
のプルトニウムの原子価を４価に調整し、溶液を冷却し
て硝酸ウラニル結晶を析出させ、母液と結晶を分離し、
結晶をウラン製品として回収する工程、分離した母液に
ついて、酸濃度及び溶液中のウランとプルトニウムの原
子価を６価に調整し、溶液を冷却して硝酸ウラニル・プ
ルトニウム結晶を析出させ、母液と結晶を分離し、分離
した結晶をウラン・プルトニウム混合製品として回収す
る工程、を具備している共晶析現象を利用した使用済核
燃料の再処理方法。2. The spent nuclear fuel is sheared and dissolved in nitric acid,
A step of removing the insoluble residue in the solution, adjusting the acid concentration and the valence of plutonium in the solution to be tetravalent, cooling the solution to precipitate uranyl nitrate crystals, separating the mother liquor and the crystals,
Step of recovering the crystals as uranium product, the separated mother liquor, the acid concentration and the valence of uranium and plutonium in the solution were adjusted to hexavalent, the solution was cooled to precipitate the uranyl-plutonium nitrate crystals, the mother liquor and the crystals And a step of recovering the separated crystals as a uranium-plutonium mixed product, a method for reprocessing spent nuclear fuel utilizing a eutecticization phenomenon.
Priority Applications (3)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP2002008320A JP2003215287A (en) | 2002-01-17 | 2002-01-17 | Reprocessing method for spent nuclear fuel using eutectic phenomenon |
| US10/262,863 US7011798B2 (en) | 2002-01-17 | 2002-10-03 | Process for reprocessing spent nuclear fuels by utilizing phenomenon of cocrystallization |
| EP20020256919 EP1329906A3 (en) | 2002-01-17 | 2002-10-04 | Process for reprocessing spent nuclear fuels by utilizing phenomenon of cocrystallization |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP2002008320A JP2003215287A (en) | 2002-01-17 | 2002-01-17 | Reprocessing method for spent nuclear fuel using eutectic phenomenon |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JP2003215287A true JP2003215287A (en) | 2003-07-30 |
Family
ID=19191412
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP2002008320A Granted JP2003215287A (en) | 2002-01-17 | 2002-01-17 | Reprocessing method for spent nuclear fuel using eutectic phenomenon |
Country Status (3)
| Country | Link |
|---|---|
| US (1) | US7011798B2 (en) |
| EP (1) | EP1329906A3 (en) |
| JP (1) | JP2003215287A (en) |
Families Citing this family (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| WO2008105928A2 (en) * | 2006-09-08 | 2008-09-04 | Michael Ernest Johnson | Process for treating compositions containing uranium and plutonium |
| RU2528399C2 (en) * | 2013-01-23 | 2014-09-20 | Федеральное государственное унитарное предприятие "Горно-химический комбинат" | Method for crystallisation separation and purification of uranyl nitrate hexahydrate and apparatus therefor |
| US9842664B2 (en) * | 2013-09-26 | 2017-12-12 | Los Alamos National Security, Llc | Recovering and recycling uranium used for production of molybdenum-99 |
| US9793023B2 (en) * | 2013-09-26 | 2017-10-17 | Los Alamos National Security, Llc | Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target |
| RU2755474C1 (en) * | 2020-12-04 | 2021-09-16 | Акционерное общество "Высокотехнологический научно-исследовательский институт неорганических материалов имени академика А.А. Бочвара" | Method for crystallization separation and purification of uranyl nitrate hexahydrate and device for its implementation |
Family Cites Families (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB863571A (en) | 1945-04-05 | 1961-03-22 | Atomic Energy Authority Uk | Separation of plutonium and fission products from uranium |
| DE3408029C2 (en) * | 1984-03-05 | 1991-06-20 | Kernforschungszentrum Karlsruhe Gmbh, 7500 Karlsruhe, De | |
| JP2551683B2 (en) * | 1990-10-01 | 1996-11-06 | 動力炉・核燃料開発事業団 | Method for separating uranium and plutonium from uranium-plutonium mixed solution |
| JP2948166B2 (en) * | 1997-04-04 | 1999-09-13 | 核燃料サイクル開発機構 | Recovery method of transuranium element from spent nuclear fuel |
-
2002
- 2002-01-17 JP JP2002008320A patent/JP2003215287A/en active Granted
- 2002-10-03 US US10/262,863 patent/US7011798B2/en not_active Expired - Fee Related
- 2002-10-04 EP EP20020256919 patent/EP1329906A3/en not_active Withdrawn
Also Published As
| Publication number | Publication date |
|---|---|
| US7011798B2 (en) | 2006-03-14 |
| EP1329906A3 (en) | 2005-01-12 |
| US20030133860A1 (en) | 2003-07-17 |
| EP1329906A2 (en) | 2003-07-23 |
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| CN110656247B (en) | Method for extracting and recovering plutonium from plutonium-containing nitric acid solution | |
| JP2001516871A (en) | Treatment of molten salt reprocessing waste | |
| JP2003215287A (en) | Reprocessing method for spent nuclear fuel using eutectic phenomenon | |
| US4759878A (en) | Process for the batch fine purification of uranium or plutonium recovered in a reprocessing process for irradiated nuclear fuel and/or fertile materials | |
| US20150228367A1 (en) | Radioactive material processing method | |
| Campbell et al. | The chemistry of fuel reprocessing: present practices, future trends | |
| JP2551683B2 (en) | Method for separating uranium and plutonium from uranium-plutonium mixed solution | |
| JP4395589B2 (en) | Method for selectively separating and recovering uranium (VI) present in aqueous solution with branched N, N-dialkylmonoamide | |
| US3148941A (en) | Dissolution of uranium-molybdenum reactor fuel elements | |
| JP3896445B2 (en) | Method for reprocessing spent nuclear fuel | |
| US10221466B2 (en) | Molybdenum recovery from aqueous nitric acid solution by solvent extraction | |
| GB2118759A (en) | Process for the recovery of plutonium from aqueous nitric solutions | |
| JP2009133671A (en) | Method of reprocessing spent fuel | |
| CN110373556A (en) | The method of radioactive element is recycled from zirconium industrial residue | |
| Kikuchi et al. | Development of reprocessing process by plutonium cocrystallization | |
| RU2743383C1 (en) | Method of processing acid-resistant uranium-containing materials | |
| JP2019015533A (en) | Method for separating minor actinoid | |
| Bartenev et al. | Recycling of vanadium alloys in fusion reactors | |
| US4756853A (en) | Process for the conversion into usable condition of actinide ions contained in the solid residue of a sulfate reprocessing process for organic, actinide-containing radioactive solid waste | |
| JP7108519B2 (en) | Isolation method for minor actinides | |
| US3560169A (en) | Manner of processing plutonium-containing uranium fuel from nuclear reactors | |
| JP3159220B2 (en) | Method for recovering uranium, plutonium and neptunium from spent nuclear fuel | |
| JPH09138296A (en) | Facility for reprocessing of spent fuel and reprocessing method | |
| JP2002022884A (en) | Recovery method for uranium | |
| Brooksbank et al. | Review of thorium fuel reprocessing experience |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| A977 | Report on retrieval |
Free format text: JAPANESE INTERMEDIATE CODE: A971007 Effective date: 20040203 |
|
| A131 | Notification of reasons for refusal |
Free format text: JAPANESE INTERMEDIATE CODE: A131 Effective date: 20040323 |
|
| A02 | Decision of refusal |
Free format text: JAPANESE INTERMEDIATE CODE: A02 Effective date: 20040615 |